method for non-destructive measurement of the activity of 238 U or U

Abstract

Method for estimating the uranium or uranium-238 activity of an object, using a spectrometric detector configured to establish a spectrum, corresponding to a histogram of the amplitude of pulses detected during a measurement period: a) positioning the detector facing the object and acquiring a measurement spectrum, the spectrum being representative of an energy distribution of β particles emitted by 238U descendants; b) from the measurement spectrum, possible formation of a corrected spectrum; c) from the measurement spectrum or the corrected spectrum, determination of a first spectral value in a first energy band extending from a minimum energy, the minimum energy being greater than or equal to 300 keV; d) application of a transfer function to the spectral value determined in step c), so as to estimate a uranium or 238U activity of the object.

Description

Title of the invention: Method for non-destructive measurement of the activity of or of U technical field

[0001] The technical field of the invention relates to a non-destructive measurement technique for activity in 238U or U. EARLIER ART

[0002] Knowledge of the radiological status of the processes and equipment of a nuclear facility is essential for establishing robust decommissioning scenarios and defining waste management, in particular its categorization and disposal. Non-destructive in-situ nuclear measurements, coupled with modeling techniques, make it possible to establish a radiological inventory of the equipment on stands, processes, and civil engineering structures.

[0003] Gamma spectrometry is one of the most commonly used passive non-destructive nuclear measurement techniques for obtaining qualitative and quantitative information on gamma-emitting radionuclides. As for pure [3] emitters, their in situ radiological characterization is made difficult by the very short range of electrons in dense media and is usually carried out in the laboratory through destructive analyses of samples collected in the field. However, these destructive laboratory measurements have certain drawbacks: questions about the representativeness of the samples collected, cost, and analysis time.

[0004] In order to address the quantification of the activity of pure [3] emitters, numerous non-destructive radiological characterization methods, based on beta counting or beta spectrometry techniques, have been developed by various laboratories worldwide to characterize, in particular, 90Sr in contaminated soils. See, for example, AV Chesnokov, SM Ignatov, VI Liksonov, VN Potapov, SB Shcherbak, LI Urutskoev, MV Ivanitskaya, A method for measuring in situ a specific soil activity of 90Sr, Nucl. Instrum. Methods Phys. Res. A 443 (2000) 197-200.

[0005] In many upstream facilities, it is necessary to characterize uranium activity, or even to determine its isotopy. Uranium-235 and Uranium-238 can be measured by gamma spectrometry. However, gamma spectrometry is quite complex to implement and relies on certain assumptions regarding the depth and gradient of contamination, particularly in porous materials such as concrete walls. These assumptions are not necessarily valid.

[0006]

[0007]

[0008]

[0009]

[0010] Furthermore, when the enrichment in 235U is high, the acquisition time can be long, due to the low intensity of gamma emissions from the 238U decay products. The inventor proposes an alternative method for estimating uranium activity in civil engineering structures, equipment, or samples taken from concrete cores. This method is based on an inexpensive and easy-to-implement device. Description of the invention A first object of the invention is a method for estimating the activity in uranium, or in uranium 238, of an object, using a detector configured to form pulses under the effect of exposure to particles [3, the detector being connected to a spectrometric measurement circuit, configured to establish a spectrum, the spectrum corresponding to a histogram of the amplitude of the pulses formed during a measurement period, the method comprising the following steps: - a) positioning of the detector facing the object and acquisition of a measurement spectrum, the measurement spectrum being representative of an energy distribution of particles [3 emitted by descendants of 238U; - b) from the measurement spectrum, possible formation of a corrected spectrum, ; - c) from the measurement spectrum or the corrected spectrum, determination of a first spectral value in a first energy band extending from a first minimum energy, the first minimum energy being greater than or equal to 300 keV; - d) application of a transfer function to the first spectral value determined in step c), so as to estimate an activity in uranium or in 238U of the object. Step b) may include the following sub-steps: - bl) interposition of a screen between the detector and the object and acquisition of a background spectrum, the screen being configured to absorb the particles [3 emitted by the object; - b2) correction of the measurement spectrum, using the background spectrum, to to obtain a corrected spectrum. The screen used in substep bl) can be an aluminium screen with a thickness greater than or equal to 3 mm or 4 mm. During step c), the first energy band can extend up to a first maximum energy, the first maximum energy being greater than or equal to 1000 keV or 1500 keV or 2000 keV. The process may include the following steps:

[0012]

[0013]

[0014]

[0015]

[0016]

[0017]

[0018]

[0019] - f) determination of a second spectral value, in a second energy band, extending between a second minimum energy and a second maximum energy, the second maximum energy being less than 300 keV; - g) estimation of an enrichment in 235U of Uranium, by comparing the first spectral value resulting from step c) and the second spectral value resulting from step f). During step a), the distance between the detector and the object can be: - greater than 1 mm or 5 mm; - and / or less than 20 mm or 10 mm. According to one embodiment, in which the object is a naturally radioactive object, the process comprises: - (i) taking into account a natural activity of the object; - (ii) modeling of a spectrum [3 of the natural activity of the object detected by the detector; - step b) involves a correction of the measured spectrum using the spectrum [3 of the natural activity of the object resulting from (ii), to obtain the corrected spectrum. Step (i) may result from a gamma spectrometry measurement performed on the object. The detector includes an organic scintillator material. The detector may include a semiconductor material. A second object of the invention is a measuring device, configured to determine the Uranium activity of an object, the device comprising: - a detector comprising a detector material configured to form pulses under the effect of exposure to particles [3; - a spectrometry circuit, connected to the detector, and configured to form a histogram of the amplitude of the pulses detected by the detector during an acquisition period; - a processing unit, connected to the spectrometry circuit, and configured to implement steps b) to d) of a process according to the first object of the invention. According to one possibility, the detector includes an organic scintillator material, the thickness of the organic scintillator material being preferably between 3 mm and 10 mm. The organic scintillator detector can extend over an area of ​​at least 20 cm on each side. The detector can be covered with an envelope configured to absorb alpha particles with energy greater than or equal to 4.5 MeV or 5 MeV. The detector may contain a semiconductor material.

[0021] The device may include a movable screen, configured to switch from a closed configuration, in which the screen is interposed between the sensing material and the object, to an open configuration in which the screen leaves a gap between the sensing material and the object.

[0022] The invention will be better understood upon reading the description of the exemplary embodiments presented later in this description, in connection with the figures listed below. FIGURES

[0023] Fig. 1 schematically represents a measuring device.

[0024] Fig. 2 shows an evolution of the energy deposited in the detector material of the device by particles [3] as a function of the thickness of the latter.

[0025] Figure 3 shows an evolution of the energy deposited in the detector material by particles [3] as a function of the thickness of an aluminum screen interposed between the detector material and the object.

[0026] Fig. 4A shows a spectrum of particle energy [3 detected by the detector material as a function of different distances between the detector material and the object, with uranium contamination assumed to be surface-bound.

[0027] Fig. 4B shows a spectrum of particle energy [3 detected by the detector material as a function of different distances between the detector material and the object, the uranium contamination being assumed to be volumetric, over a thickness of 2 mm.

[0028] Figure 5 shows particle energy spectra [3] detected by the detector material in the presence of an aluminum screen, and without the aluminum screen. A corrected spectrum resulting from the subtraction of the spectra modeled with and without the screen is also shown.

[0029] Fig. 6 shows a spectrum of particle energy [3 detected by the detector material due to the natural radioactivity of the object, as well as spectra of particle energy [3 detected by the detector material with and without subtraction of the spectrum of natural activity.

[0030] Fig. 7 shows simulations of particle energy spectra [3 detected by the detector material in different configurations.

[0031] Fig. 8A shows simulations of particle energy spectra [3 detected by the detector material resulting from different radioelements, for a low enrichment of 235U.

[0032] Fig. 8B shows simulations of particle energy spectra [3 detected by the detector material resulting from different radioelements, for a high enrichment of 235U.

[0033] Fig. 9 shows an energy spectrum of a particles detected by the detector material.

[0034] Fig. 10 shows a spectrum of particle energy [3 detected by the detector material.

[0035] Fig. 11 shows the evolution of a transfer function, allowing a spectral value to be converted into an activity level, and this for different depths of contamination.

[0036] Figures 12A, 12B and 12C show experimental mappings carried out with a device according to the invention.

[0037] Fig. 13A shows the evolution of a ratio between a first spectral value and a second spectral value as a function of the enrichment in 235U, taking into account surface contamination.

[0038] Fig. 13B shows the evolution of a ratio between a first spectral value and a second spectral value as a function of the enrichment in 235U, taking into account volumetric contamination.

[0039] Fig. 14 shows the main steps of a method for determining U-activity by implementing the invention. PRESENTATION OF SPECIFIC IMPLEMENTATION METHODS

[0040] Figure 1 represents a measuring device for measuring the uranium activity of an object 2. The device includes a scintillator detector 10, comprising a scintillator material 11, preferably organic, preferably based on polyvinyltoluene (PVT). Light pulses are formed through interactions between ionizing radiation and the scintillator material. These light pulses are converted into electrical pulses by one or more photodetectors 12. The electrical pulses are then processed by a spectrometry circuit. The spectrometry circuit 13 is configured to form an amplitude histogram of the pulses detected by the organic scintillator during an acquisition period.

[0041] By implementing an energy calibration function, resulting from an energy calibration, it is common to establish a correspondence between the pulse amplitude and energy values. When an ionizing particle deposits all its energy in the scintillator, the amplitude of the pulse it generates corresponds to the particle's energy before interaction in the detector material.

[0042] The use of an organic scintillator detector is suitable for performing charged particle spectrometry, of type [3]. An organic scintillator is also sensitive to ionizing X-ray or gamma-ray photons. However, the materials forming an organic scintillator have a low atomic number, which makes them unsuitable for the formation of photoelectric interactions. Thus, an organic scintillator is considered unsuitable for X-ray or gamma-ray spectrometry applications.

[0043] The scintillator detector is covered by an optically sealed casing 14, of A thin layer, for example an aluminized PET (Polyethylene Terephthalate) film, 18 pm thick, is used to ensure airtightness against ambient light. The thin layer minimizes radiation attenuation [3]. In the following text, the term [3] refers to a [3] particle. The envelope thickness can reach 30 pm to stop alpha particles with energies of 4.5 MeV or 5 MeV.

[0044] The thickness e of the scintillator detector is 4 mm. Details regarding the scintillator thickness will be given in relation to [Fig. 2]. An organic scintillator material has the advantage of being relatively insensitive to gamma radiation due to its low atomic number. Furthermore, this type of scintillator limits the backscattering of particles [3]. Another advantage is the stability of the scintillator material's response to temperature variations. The response, in terms of the light intensity produced, when exposed to the same radiation, is stable from 0° to 50°C, which is suitable for field use.

[0045] The measuring device includes a processing unit 20, configured to implement spectrum processing steps described below. The processing unit 20 is programmed to execute instructions encoded in a memory, connected to the processing unit by a wired or wireless link. The processing unit 20 may, in particular, include a microprocessor.

[0046] According to one variant, the detector may comprise a semiconductor material, suitable for spectrometry [3. It may for example be a silicon-type semiconductor, for example planar silicon.

[0047] The advantage of an organic scintillator material is its ability to be manufactured in different dimensions and shapes. When the object is a sample taken from it, the shape can be adapted to the shape of the sample.

[0048] The detector 10 includes a movable screen 15, acting as a shutter, configured to be positioned: - in a closed position, between detector 10 and the object to be measured 2; - or in an open position, freeing up the space between detector 10 and object 2, as shown in [Fig.1].

[0049] In this example, the device is primarily dedicated to the radiological inspection of walls. In order to be deployed along large surfaces, the surface area of ​​the detector material, in a plane perpendicular to its thickness, is 50 cm x 50 cm. This allows a large area to be covered with each measurement, while remaining sufficiently compact and lightweight for easy handling.

[0050] In this example, the screen 15 is movable in translation in a plane parallel to the detector material 11. The screen is, for example, made of aluminum, preferably with a thickness of 4 mm. Details relating to the thickness of the screen 15 are given in connection with [Fig. 3]. The screen 15 can be connected to the detector material by a system a sliding mechanism allows the screen to be translated relative to the detector material.

[0051] Object 2 is an object to be monitored, capable of exhibiting a mass or surface activity of Uranium, the isotope being known or unknown. In the examples described below, the object is a concrete wall.

[0052] The invention is based on the detection, by the detector 10, of [3] particles emitted by decay products of Uranium isotopes. These include 234mPa, which is in radioactive equilibrium with 238U, and emits [3] particles up to a maximum energy E[3max] of 2269 keV. The detection of [3] particles emitted by 231Th (a decay product of 235U), with E[3max] = 308 keV, or 234Th (a decay product of 238U), with E[3max] = 198 keV, can also be exploited.

[0053] Figure 2 represents the quantity of energy (ordinate axis - unit MeV.Bq 1 .g) deposited by [3] particles emitted by a homogeneous distribution of 234 mPa in a concrete depth of 2 mm. The abscissa axis corresponds to the thickness (unit mm) of the scintillator material 10. The depth of 2 mm corresponds to the maximum path length of the [3] particles with energy of 2269 keV (maximum emission energy of 234 mPa) in the concrete. Figure 2 was obtained by simulation, considering a distance d = 7.5 mm between the detector material and the concrete. It can be observed that, using a thickness e of 4 mm, the scintillator material absorbs 97% of the [3] emission energy of 234 mPa. The simulation was performed using the MCNP6 (Monte Carlo N Particles) calculation code, which is a reference code in the field of modeling interactions between ionizing radiation (|3, y, neutrons) and matter.

[0054] Figure 3 represents the quantity of energy (ordinate axis - unit MeV.Bq 1 .g) deposited by [3] particles emitted by a homogeneous distribution of 234 mPa in a concrete depth of 2 mm. The abscissa axis corresponds to the thickness (unit mm) of the aluminum screen 15. Figure 3 was obtained by simulation, considering a distance d = 7.5 mm between the detector material and the concrete. It can be observed that beyond a thickness e of 4 mm of aluminum, the gain in terms of absorption of [3] particles is negligible. A thickness of 4 mm is therefore sufficient to effectively block the detector material from [3] particles emitted by 234 mPa.

[0055] Figures 4A and 4B address the issue of the distance between the detector and the object to be monitored. Since the monitored object is potentially contaminating, it is preferable to avoid direct contact between the detector 10 and the object. A slight offset is useful to allow the screen 15 to move from the open to the closed position. [Fig. 4A] shows simulations of the spectra of the energy deposited by particles [3] in the scintillator material, as described in connection with [Fig. 1]. A homogeneous surface distribution of 234 mPa on a wall was considered, and several distances d from the detector to the wall were taken into account: d = 2 mm; d = 7 mm; and d = 10 mm. In [Fig. 4A], the x-axis represents the energy (MeV) and the y-axis represents the The ordinate corresponds to the count rate (s * Bq * cm²). It is observed that between 2 mm and 10 mm, the spectrum is not significantly modified. The difference between the spectra acquired at distances of 7 mm and 10 mm is negligible.

[0056] Figure 4B shows simulations of the same type as those described in connection with Figure 4A. In Figure 4B, a homogeneous volumetric pressure distribution of 234 mPa was considered over a 2 mm thickness of concrete. The same distances d from the detector to the wall were used as in Figure 4A. The difference between the spectra acquired considering distances of 2.5 mm and 10 mm is negligible.

[0057] The inventor considers the optimal distance between object 2 and detector 10 to be between 5 mm and 10 mm. Beyond 10 mm, there is a risk of reduced efficiency, i.e., a decrease in the number of particles [3] detected. Below 5 mm, handling the screen becomes difficult, as it is too close to the object, with a risk of contact between the screen and the object. Therefore, the distance between the detector and the object is 7.5 mm.

[0058] Figure 5 shows the effect of blocking the detector with the screen 15 on the spectrum measured by the detector. Figure 5 represents spectra measured experimentally in front of a uranium-contaminated wall. A spectrum without the screen (screen in the open position, as shown in Figure 1) and a spectrum with the screen (screen in the closed position) were measured, respectively. The spectra were measured with the same acquisition time of 120 seconds. The screen was a 4 mm thick aluminum plate. The spectrum with the screen, denoted Spbdf, is considered representative of the background noise γ. The background noise γ is due to γ-emitting radioelements present in the object (e.g., natural radioelements) or in the detector environment. Subtracting the spectrum with the screen, Spbdf, from the spectrum without the screen yields a spectrum corrected for the background noise γ.The corrected Sp[3 spectrum, or raw [3 spectrum, is considered to be solely representative of the [3 particles emitted by the object.

[0059] Among the naturally occurring radioelements potentially present in the object are, for example, 40K, as well as decay products of 232Th and 238U. Such radioelements are, for example, present in concrete objects. Background noise may also result from artificial radioactivity present in the detector's environment.

[0060] When the object contains natural radioactivity, the raw [3] spectrum Sp[3 includes a natural component due to [3] particles emitted by natural radioelements. This natural component can, as a first approximation, be neglected because it is generally quite small. Table 1 shows the various [3] emitting radionuclides of natural and artificial origin likely to be found in a concrete wall potentially contaminated with uranium.

[0061] [Tables 1] £mé ttëurjbèta Energy intensity Origin 4473⁄4 4- — -À 1 ■ 0 f â A. 7 A Natural Decay chain of F235U 4. 4 ' s 3 177, SA: 7$ Natural & Artmdei 4.4 ? Ac AA S 5 N st u re 1 22 3Fr i JA ïÿ .A 2 Natural §i.2â Natural :: 1412 Natural Chain of desmtê grâtiôn de i'23S U ' : - r TS'îs 77. g Nature 4 Art > e 7 HS •. P <!• 47 Naturel & Amusées ^2-H.' le lë 92 Natures 214 b; 3270 le4' Naturel 210 P b & 3 . B À 4 A Naturel 21W 1112 4 'Jbj Naturel Chaîne de dêsintégràdôn du 232Th 2288a 3$ A •1'7 Naturel Naturel 112P& S69.9 4 4: : 5 Naturel 212È1 2232.1 55.31 N à turel iÛST: 18C1.5 4^:7 Naturel

[0062] Table 1

[0063] Figure 6 shows the corrected [3] spectrum Sp[3], as described in connection with Figure 5 (raw [3] spectrum), as well as a [3] spectrum, also corrected for background noise y, measured facing an uncontaminated concrete wall (natural background noise spectrum or background noise spectrum RN, denoted SpRN). The uncontaminated concrete wall was located in the same facility as the contaminated wall, on which the corrected [3] spectrum was formed. Thus, the [3] background noise spectrum RN, denoted SpRN, can be considered representative of a natural [3] contribution to the corrected [3] spectrum. Figure 6 shows a net [3] spectrum, denoted Sp[3'], obtained by subtracting the natural background noise [3] spectrum SpRN from the raw [3] spectrum. The net [3] spectrum Sp[3' is thus representative of the artificial [3] activity in the measured wall.

[0064] Figure 7 shows simulated [3] spectra using the MCNP6 calculation code, in a detector material as previously described (50 cm x 50 cm x 4 mm), taking into account the specific activities of 232Th, 238U, and 40K measured by gamma spectrometry on a wall considered uncontaminated. The gamma spectrometry measurement allowed for the estimation of activity levels of 15 Bq / kg for 232Th, 22 Bq / kg for 238U, and 485 Bq / kg for 40K. These specific activities allow for the simulation of the natural component in the [3] spectrum measured by the detector. In Figure 7, we have re- presented: - the component due to 232Th in the spectrum [3; - the component due to 238U in the spectrum [3; - the component due to 40K in the spectrum [3; - the component due to 235U in the spectrum [3, assuming an enrichment natural uranium: the mass of 235U corresponds to 0.72% of the mass of uranium;

[0065] In [Fig.7], a simulation of the spectrum [3 resulting from 234m is also shown Pa taking into account a homogeneous surface activity of 0.4 Bq.cm2 and an enrichment of 1% by mass in 235U (UB spectrum), and taking into account an enrichment of 8.16% by mass in 235U (UH spectrum).

[0066] In these simulations, the distance between the wall and the detector was considered to be 7.5 mm, and the density of the concrete was considered to be 2.3 g.cm3.

[0067] It is observed that: - the natural component of the [3] spectrum is dominated by 40K, up to the energy 1311 keV, which corresponds to the maximum energy of the [3] particles emitted by 40K; - the measurement of the spectrum [3 resulting from 234mPa is largely greater than the natural contribution at low enrichment as well as at high enrichment, beyond 1 MeV. - It is possible to model the natural component in a [3] spectrum measured by the detector, taking into account the mass activities, established a priori or measured, of the main natural [3] emitting radioelements.

[0068] Furthermore, simulations have shown that the y component of the natural activity is not significantly absorbed by the aluminum screen. Therefore, subtracting the background noise spectrum, measured in the presence of the aluminum screen, eliminates the y component of the natural activity.

[0069] Figures 8A and 8B show simulations of spectra [3] measured by a detector as described in [Fig. 1], taking into account a homogeneous surface activity of uranium, on a wall element with a side of 50 cm, taking into account a 235U enrichment of 1% and 82.55% by mass, respectively. Surface activity is understood to mean an activity distributed over a depth of 10 pm. In each of these figures, the contributions of 234mPa, 234Th, and 231Th are shown. 234mPa and 234Th are 238U decay products, while 231Th is a 235U decay product. The distance between the detector and the wall was considered to be 7.5 mm. On each of these figures, the ordinate axis corresponds to a counting rate (s '.Bq '.cm2) and the abscissa axis corresponds to an energy level (MeV).

[0070] Figures 8A and 8B show that regardless of the enrichment level, the The spectrum is dominated by the contribution of 234 mPa, which is weaker at high enrichment ([Fig. 8B]) than at low enrichment ([Fig. 8A]). It is observed that by considering a first energy range AE1 delimited by a first minimum energy Elmin, greater than or equal to 300 keV, and by summing the spectrum over this first energy range AE1, the spectral value obtained depends only on the activity of 234 mPa. By spectral value, we mean a value measured from the spectrum. This can notably be a count rate or a count within the energy range. The spectral value obtained over the first energy range AE1 is a first spectral value denoted NI.

[0071] Preferably, the first energy range AE1 extends up to a first maximum energy Elmax greater than or equal to 2269 keV, the latter corresponding to the maximum emission energy of a [3] particle at 234 mPa. For example, the first energy range AE1 is [300 keV; 2500 keV]. Such an energy range makes it possible to collect the entire useful spectral content of the [3] activity at 234 mPa. The first maximum energy Elmax can be equal to 2300 keV, or be less than 2000 keV, 1500 keV, or 1000 keV. It is noted that the first maximum energy Elmax is preferably less than or equal to 2500 keV, or even 2300 keV. Thus, although the first optimal energy range is [300 keV; 2500 keV], a narrower energy range within this interval may be suitable. An energy range extending from 300 keV, or from a first minimum Elmin value greater than or equal to 300 keV, and less than or equal to 1000 keV, or 1500 keV, may be suitable.It is noted that beyond 1000 keV, the spectral channel values ​​decrease. Therefore, it is preferable for the first minimum energy Elmin to be between 300 keV and 1000 keV. It is possible to disregard a first maximum energy: Elmax = + , the range considered is then [Elmin, + ], with 300 keV < Elmin < 1000 keV.

[0072] The first spectral value NI, in the first energy range AE1 = [Elmin, Elmax] defined previously, is then processed by a transfer function, so as to be able to estimate an activity level in 238U. The establishment of the transfer function is described subsequently, in connection with [Fig. 11]. Taking into account the enrichment q then allows the conversion of the 238U activity into U activity.

[0073] Figures 8A and 8B show another interesting aspect of the invention: a second energy range AE2, between 0 keV and 300 keV, can be defined in which the spectrum [3] is influenced by 231Th, this decay product of 235U. Thus, a second energy range AE2=[E2min, E2max], between a second minimum energy E2min>0 and a second maximum energy E2max<300 keV, can be defined in which the spectral value N2, called the second spectral value, depends on 235U, particularly at high enrichment.

[0074] A ratio N1 / N2 of the first and second spectral values ​​depends on the enrichment r|. Thus, the experimental measurement of this ratio can be used to estimate the enrichment q of uranium.

[0075] When the enrichment p of uranium in 235U is unknown, the first spectral value NI allows the 238U to be estimated using the transfer function FT. Indeed, the spectral content in the first energy range does not depend on the enrichment. The ratio of the first and second spectral values ​​allows the enrichment p to be estimated. The 238U activity can then be used to estimate the U activity. The 235U enrichment can also be determined experimentally by gamma spectrometry.

[0076] Uranium contains alpha-emitting isotopes, primarily 234U or 238U. To estimate the contribution of alpha particles to the measured spectrum [3], a 244Cm source with an activity of 2800 Bq was placed 3 mm from the detector material. 244Cm emits alpha particles with energies above 5700 keV, which is much higher than the energies of alpha particles emitted by U isotopes, for example, 4775 keV for 234U. Figure [9] shows the spectrum measured over an acquisition period of 900 s. It can be observed that the contribution of alpha particles is limited to low energies, below 270 keV. Thus, the spectral value in the first energy range can be considered unaffected by alpha particles emitted by uranium isotopes.

[0077] The alpha particles detected by the detector may be likely to influence the spectral value N2 in the second energy band AE2. This is particularly the case when the detector includes a semiconductor material, for example Si. In order to avoid an influence of alpha particles, the envelope 14 is advantageously dimensioned to stop alpha particles with an energy of 4.5 MeV or 5 MeV. When the envelope is made of aluminized PET (Polyethylene Terephthalate), a thickness of 30 pm makes it possible to stop alpha particles, while absorbing, in a way considered negligible, the particles [3.

[0078] On [Fig.9], the few shots observed at an energy greater than 270 keV correspond to y photons after subtraction of background noise, resulting from statistical fluctuations.

[0079] Figure 10 shows a spectrum [3] measured experimentally on a concrete wall contaminated with 1% enriched uranium. The first energy range, between 300 keV and 2700 keV, is shown on this spectrum. The detector was an EJ200 detector (Eljen Technology), with a sensitive area of ​​2430 cm² (493 mm on each side). Energy calibration of the spectrum, i.e., the correlation between the pulse amplitude values ​​and the energy, was performed using a 207Bi source, which emits electrons at discrete energy values ​​by internal conversion.

[0080] In order to convert the spectral value, in the first energy range, into the activity value of 238U, it is necessary to take into account a transfer function. The The transfer function can be obtained by modeling spectra [3] corresponding to known activities at 238U, and therefore at 234 mPa. Different transfer functions can be determined, corresponding respectively to different assumptions about the distribution of contamination within the analyzed object. Uranium contamination can be considered as surface contamination (for example, on metallic objects), or as diffusion within the object, for example, in the case of liquid contamination of porous objects, such as soil or a concrete wall.

[0081] Figure 11 shows the result of numerical modeling of a detector as previously described, positioned at a distance of 7.5 mm from a concrete wall with a density of 2.3, whose uranium-238U activity was distributed over different depths. A unit activity of 238U was considered (1 Bq / g or 1 Bq / cm² assuming surface contamination). The detector model was validated in the laboratory by comparing spectra [3] respectively measured and modeled by exposing the detector to a standard 207Bi source.

[0082] In the case of surface contamination, an activity distributed over the first 10 pm of the wall was modeled. The transfer function is 1050s '.Bq '.cm2 ± 15%.

[0083] In the case of volumetric contamination, the maximum path length in the concrete of the particles [3 emitted by 234 mPa, which can be estimated at 2 mm, was taken into account. Given this small thickness, the contamination gradient was considered to be homogeneous.

[0084] In Figure 11, the ordinate axis corresponds to the value of the transfer function FT (unit s'.Bq'.cm2) and the abscissa axis corresponds to the considered contamination depth. In Figure 11, the transfer function corresponds to the first spectral value N1, in an energy range of [300 keV - 2500 keV], for a unit activity of IBq.g'.

[0085] If NI corresponds to the counting rate in the first energy band AE1, the activity A[238t7], in Bq.g 1 or in Bq.cm2, is obtained according to the expression:

[0086] ^2381 / ] = (1)

[0087] The detector described above was implemented on a gaseous diffusion uranium enrichment plant. The detection limit was such that:

[0088] jn _ kHr+2xSD (2) — FTxTs

[0089] With

[0090] k\.p — 1.96: risk level of type fi for a 95% confidence interval;

[0091] Ts; acquisition time, in seconds;

[0092] FT = transfer function, expressed in Bq.g or in Bq.cm2

[0093] SD is the detection threshold, such that [00941 so=^,x®

[0095] k-a ~ 1 -96: type a risk level for a 95% confidence interval;

[0096] n: ratio of the respective acquisition times of background noise and measurement

[0097] ; count rate measured in the first energy band AE1, in a Background noise measurement, on a wall considered uncontaminated. Unit: counts per second.

[0098] Depending on the unit of the transfer function, the detection limit is expressed in Bq.g or in Bq.cm2

[0099] Tables 2 and 3 show detection limit values ​​expressed as surface activity and mass activity, respectively. The values ​​are calculated from equations (2) and (3), assuming a measurement duration equal to the background noise measurement duration. A contribution of natural activity to the spectrum, described in relation to [Fig. 7], was also taken into account. The natural isotope activities considered were 15 Bq / kg for 232Th, 22 Bq / kg for 238U, and 485 Bq / kg for 40K. Each measurement includes one measurement with the aluminum screen and one measurement without the aluminum screen, of equal duration.

[0100] [Tables2] LD in Bq.cm-2 . , ,, FT Ts ]s) Radionuclide ROI r60f [sq . , „ . (s LBqhcm2) go 180 300 600 1800 >300- W^Pa) ' , ..48 + 8% 1050 + 15%. 4.81O'3 3.4.10-® 2.2.10- 1.5.10-® 8.7.10-4 ' ' 2500^kev

[0101] Table 2

[0102] [Tables3] LD in Bq.g1 Radionuclide ROI Fsof (s: ! FT (sLBqLg} 60 180 Ts (s] 300 600 1800 48+8 % 149+15% 3.4.10-2 2.4.10-2 1.5.10-2 1.1.10-2 6.1.10-3

[0103] Table 3

[0104] It is observed that the method makes it possible to obtain low detection limits in reasonable acquisition times. The detection limits presented in Tables 2 and 3 concern 238U via the detection of the [3] spectrum at 234 mPa. The higher the enrichment, the lower the quantity of 238U relative to the quantity total uranium. Also, if we want to maintain a low uranium detection limit, we need to increase the acquisition time.

[0105] Mapping was carried out on the walls of the gaseous diffusion uranium enrichment facility. [Fig. 12A] shows measured count rates (cps = counts per second) on surface elements measuring 50 cm x 50 cm. [Fig. 12B] shows the corresponding 238U activity levels. Since the 235U enrichment is known, U activity levels were estimated and are shown in [Fig. 12C]. When the enrichment is not known, it can be determined by conventional methods, for example, by gamma spectrometry.

[0106] As previously described, in connection with Figures 8A and 8B, the enrichment q in 235U can be evaluated from a ratio between the first spectral value NI and the second spectral value N2. The second energy band AE2 is between 0 keV and 250 keV. For example, a second energy band AE2 extending from E2min = 150 keV to E2max = 300 keV can be considered. The second minimum limit E2min of 150 keV was determined in such a way as to limit the influence of gamma radiation: between 150 keV and 300 keV, the contribution of gamma radiation to the spectrum [3] was found to be stable and can be easily subtracted by taking into account the measurement with the aluminum screen. By taking into account a second minimum energy E2min less than 150 keV, the contribution of gamma radiation in the spectrum [3] can fluctuate more, due to the higher sensitivity of the plastic scintillator material to low energy photons.

[0107] Figure 13A shows the evolution of a ratio between the first and second spectral values. The first energy band is [300 keV - 2500 keV] and the second energy band is [150 keV - 300 keV]. The contamination is assumed to be surface-bound, i.e., considered to be distributed over a thickness of 10 pm.

[0108] Fig. 13B shows the evolution of a ratio between spectral values ​​(count rate) in the first energy band [300 keV - 2500 keV] and in the second energy band [150 keV - 300 keV] in the case of mass contamination, distributed over a thickness of 2 mm.

[0109] In Figures 13A and 13B, the y-axis represents the N1 / N2 ratio and the x-axis represents the 235U enrichment, expressed as a percentage. Figures 13A and 13B were created based on modeling using the MCNP6 calculation code.

[0110] A correlation is observed between the enrichment and the N1 / N2 ratio. Thus, by applying an enrichment calibration function, as described in connection with Figures 13A or 13B, the enrichment can be estimated from the N1 / N2 ratio. Figures 13A and 13B show the establishment of enrichment calibration functions, respectively for surface activity and volumetric activity.

[0111] It is specified that the use of spectrometry [3 to estimate an enrichment rate q may constitute an independent aspect of the invention.

[0112] In practice, the use of a background noise spectrum, obtained according to the closed configuration with the screen 15, is not necessary. Indeed, if a rough check is desired, measurements carried out solely according to the open configuration, i.e., without the screen, may prove sufficient for a first-level check. Coupling between the open and closed configurations makes it possible to subtract the background noise, mainly due to gamma photons, which allows for more precise measurements: a more accurate quantification of the 238U activity or a more precise enrichment value.

[0113] Furthermore, experience has shown that in the first energy band AEl, extending from 300 keV to 2500 keV, the contribution of background gamma noise due to natural radioactivity is stable in the same installation. However, performing measurements with and without a screen increases the total measurement time. Given the low dispersion of the contribution, it is conceivable not to perform a systematic measurement in the closed configuration. It is possible to subtract a stored spectrum, obtained in the closed configuration, from a spectrum measured in the open configuration.

[0114] Figure 14 summarizes the main steps of a process according to the invention.

[0115] Step 100: Positioning a detector facing the object to be characterized

[0116] Step 110: Acquisition of a spectrum, in the open configuration: this is the measured spectrum Sp.

[0117] Step 120: acquisition of a background noise spectrum Spbdf, in the closed configuration, or taking into account a spectrum acquired in the closed configuration.

[0118] Step 130: subtraction of the spectrum resulting from step 120 from the spectrum resulting from step 110. The spectrum resulting from this step is a raw [3 Sp[3] spectrum.

[0119] Step 140: Taking into account the natural activity [3 of the object. This step includes the following sub-steps: - Sub-step 141: taking into account a level of natural activity of the object: the level of natural activity may have been estimated by a measurement on a comparable object, considered to be uncontaminated. - Substep 142: estimation of a [3 natural SpRN spectrum (or [3 RN] spectrum) resulting from step 141.

[0120] Alternatively, step 140 includes a substep 143 for measuring a raw [3] spectrum on an object considered representative of the object being measured, and uncontaminated. This involves carrying out steps 110 and 120 on the object considered representative, which makes it possible to obtain the natural [3] spectrum

[0121] Step 140 is preferable, but optional.

[0122] Step 150: Correction of the raw [3] spectrum, resulting from step 130, by taking into account the natural [3] spectrum resulting from step 140. This step is optional. Step 150 makes it possible to obtain a net [3] spectrum (or corrected [3] spectrum) denoted Sp[3'

[0123] Step 160: determination of a first spectral value NI in the first energy band AE1 from the [3 net Sp[3' spectrum resulting from step 150 or the [3 gross spectrum resulting from step 130.

[0124] Step 170: application of the transfer function FT to the first spectral value so as to estimate an activity of the object in 238U denoted A[238U]

[0125] Step 180: taking into account a mass enrichment rate q of 235U. The enrichment is either known, as it depends on the installation, or unknown, in which case it may result from a non-destructive measurement such as gamma spectrometry. The enrichment can also be determined from the spectrum from the corrected [3] spectrum resulting from step 150 or from the net [3] spectrum resulting from step 130. For this, step 190 is implemented.

[0126] Step 190: Step 190 comprises the following sub-steps: - Substep 191: determination of a second spectral value N2 in the second energy band AE2 from the net [3] spectrum resulting from step 150 or from the gross [3] spectrum resulting from step 130. - Substep 192: application of the enrichment calibration function to the N1 / N2 ratio to estimate an enrichment q. The enrichment calibration function depends on assumptions about the nature of the activity: surface activity (see [Fig.l3A]) or volumetric activity (see [Fig.l3B]).

[0127] Step 200: from the enrichment q, resulting from step 180, and the activity of 238U resulting from step 170, determination of the uranium activity A[U].

[0128] In some applications, the activity in 238U resulting from step 170 is sufficient: steps 180 to 200 are not implemented.

[0129] Steps 130 to 200 can be implemented by the processing unit.

[0130] The invention can be implemented on fuel enrichment or manufacturing facilities.

Claims

Claims

1. Method for estimating the uranium, or uranium 238, activity of an object (2), using a detector (10) configured to form pulses under the effect of exposure to particles [3, the detector being connected to a spectrometric measurement circuit (13), configured to establish a spectrum, the spectrum corresponding to a histogram of the amplitude of the pulses formed during a measurement period, the method comprising the following steps: - a) arrangement of the detector facing the object and acquisition of a measurement spectrum (Sp), the measurement spectrum being representative of an energy distribution of particles [3 emitted by 238U descendants; - b) from the measurement spectrum, possible formation of a corrected spectrum (Sp[3 ; Sp[3');- c) from the measurement spectrum or the corrected spectrum, determination of a first spectral value (NI) in a first energy band (AE1) extending from a first minimum energy (Elmin), the first minimum energy being greater than or equal to 300 keV; - d) application of a transfer function (FT) to the first spectral value determined during step c), so as to estimate a uranium or 238U activity of the object.;

2. Method according to claim 1, in which step b) comprises the sub-steps: - bl) interposition of a screen (15) between the detector and the object and acquisition of a background spectrum, the screen being configured to absorb the particles [3 emitted by the object; - b2) correction of the measurement spectrum, using the background spectrum, to obtain a corrected spectrum.

3. Method according to claim 2, in which the screen (15) used during sub-step bl) is an aluminum screen whose thickness is greater than or equal to 3 mm or 4 mm.

4. A method according to claim any one of the preceding claims, wherein in step c), the first energy band extends up to a maximum energy (Elmax), the maximum energy being greater than or equal to 1000 keV or 1500 keV or 2000 keV.

5. Method according to any one of the preceding claims, comprising the steps: - f) determining a second spectral value (N2), in a second energy band (AE1), extending between a second minimum energy (E2min) and a second maximum energy (E2max), the second maximum energy being less than 300 keV; - g) estimating an enrichment in 235U of the Uranium, by comparing the first spectral value resulting from step c) and the second spectral value resulting from step f).

6. Method according to any one of the preceding claims, wherein during step a), the distance between the detector and the object is: - greater than 1 mm or 5 mm; - and / or less than 20 mm or 10 mm.

7. Method according to any one of the preceding claims, in which the object is a naturally radioactive object, the method comprising: - (i) taking into account a natural activity of the object; - (ii) modeling a spectrum [3 of the natural activity (SpRN) of the object detected by the detector; - step b) comprises a correction of the measured spectrum using the spectrum [3 of the natural activity of the object resulting from (ii), to obtain the corrected spectrum.

8. A method according to claim 7, wherein step (i) results from a gamma spectrometry measurement carried out on the object.

9. A method according to any preceding claim, wherein the detector comprises an organic scintillator material.

10. A method according to any one of claims 1 to 9, wherein the detector comprises a semiconductor material.

11. Measuring device, configured to determine a uranium activity of an object (2), the device comprising: - a detector (10), comprising a detector material (11) configured to form pulses under the effect of exposure to particles [3; - a spectrometry circuit (13), connected to the detector, and configured to form a histogram of the amplitude of the pulses detected by the detector during an acquisition period; - a processing unit (20), connected to the spectrometry circuit, and configured to implement steps b) to d) of a method according to any one of the preceding claims.

12. Device according to claim 10, in which the detector comprises an organic scintillator material (11), the thickness of the organic scintillator material being between 3 mm and 10 mm.

13. A device according to claim 11, wherein the detector comprises a semiconductor material (11).

14. Device according to any one of claims 11 to 13, in which the detector is covered with an envelope (14) configured to absorb particles α with energy greater than or equal to 4.5 MeV or 5 MeV.

15. Device according to any one of claims 10 to 14, in which the device comprises a movable screen (15), configured to pass from a closed configuration, in which the screen is interposed between the detector material and the object, to an open configuration in which the screen leaves a space free between the detector material and the object.

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